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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 379 records · Page 21
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Enhanced Performance of Pt Nanoparticles on Ni-N Co-Doped Graphitized Carbon for Oxygen Reduction Reaction in Polymer Electrolyte Membrane Fuel Cells

Since the reaction rate and cost for cathodic catalyst in polymer electrolyte membrane fuel cells are obstacles for commercialization, the high-performance catalyst for oxygen reduction reaction is necessary. The Ni encapsulated with N-doped graphitic carbon (Ni@NGC) prepared with ethylenediamine and carbon black is employed as an efficient support for the oxygen reduction reaction. Characterizations show that the Ni@NGC has a large surface area and mesoporous structure that is suitable to the support for the Pt catalyst. The catalyst structure is identified and the size of Pt nanoparticles distributed in the narrow range of 2–3 nm. Four different nitrogen species are doped properly into graphitic carbon structure. The Pt/Ni@NGC shows higher performance than the commercial Pt/C catalyst in an acidic electrolyte. The mass activity of the Pt/Ni@NGC in fuel cell tests exhibits over 1.5 times higher than that of commercial Pt/C catalyst. The Pt/Ni@NGC catalyst at low Pt loading exhibits 47% higher maximum power density than the Pt/C catalyst under H2-air atmosphere. These results indicate that the Ni@NGC as a support is significantly beneficial to improving activity.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2022

DEVELOPMENT OF AN APPROACH FOR DOWNBLENDING AND DISPOSITION OF IRRADIATED GRAPHITE FUEL

The IGR reactor, a pulse-type test experimental reactor, reached full power in 1961. The reactor contains a center axial experimental ampoule that is subjected to a large pulse of neutrons (neutron flux maximum density of 7×1016 n/cm2·s, peak power 10GW) when the reactor is operated. The fuel form for the IGR reactor is graphite cubes and rods impregnated with highly enriched (90%) uranyl-dinitrate (UO2(NO3)2·6H2O). This fuel is eligible for return to Russia for reprocessing under the Russian Research Reactor Fuel Return program. However due to some technical and licensing issues it was decided to immobilize this fuel in Kazakhstan. The disposition approach thus selected for the spent HEU fuel involves down-blending of the fuel followed by cementation. Development of the process for down-blending and subsequent cementation requires a process that satisfies many criteria, including fissile content, compressive strength, permeability, and containment of radioactivity, among others. One key goal of the final product is to meet the criteria for termination of safeguards. This paper will describe the overall challenge associated with disposition of irradiated HEU graphite spent fuel, the identification of the objectives of the disposition process, and the constraints associated with development of the process and the waste form.

Gaines, Kris↗
DOE OSTI2022

Methods and systems for microwave assisted production of graphitic materials

Systems and methods for plasma based synthesis of graphitic materials. The system includes a plasma forming zone configured to generate a plasma from radio-frequency radiation, an interface element configured to transmit the plasma from the plasma forming zone to a reaction zone, and the reaction zone configured to receive the plasma. The reaction zone is further configured to receive feedstock material comprising a carbon containing species, and convert the feedstock material to a product comprising the graphitic materials in presence of the plasma.

Zeller, Kurt W.↗
DOE OSTI2022

Session 3: Irradiation and Thermal Stesses - NRC Graphite Behavior Model

Graphite anisotropic microstructure Graphite crystal structure and basal planes Response to irradiation and temperature C-axis and a-axis behavior The effect of porosity Nano-and micro-length scale Microscopic to macroscopic response Material property changes Defect length-scale factors Internal stress build-up Irradiation induced dimensional change Temperature effects Stress increases & property changes

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Preliminary Benchmark Uncertainties for Deimos, a HALEU-Fueled and Graphite-Moderated Advanced Reactor Testbed

Many advanced reactor concepts will make use of various uranium fuels with levels of enrichment higher than previously seen in current light water reactors. In particular, High-Assay Low Enriched Uranium (HALEU), that is uranium enriched to 235 U ≈ 20 w/o%, is planned to be used in over ten new reactor concepts. HALEU is attractive for advanced reactors as it enables longer intervals between refueling. Unfortunately, little to no experience with HALEU is available in experimental literature raising concerns for not only licensing advanced re actors but also fabrication and transportation of HALEU fuels. This is where Deimos, a Los Alamos National Laboratory internal project, comes in. Deimos is a new critical experiment scheduled for FY24 at the National Criticality Experiments Research Center (NCERC). Deimos is a graphite moderated, graphite and beryllium reflected critical experiment making use of HALEU TRi-structural ISOtropic (TRISO) fuel from the Compact Nuclear Power System (CNPS). This transaction entails a brief description of efforts to benchmark Deimos for inclusion into the International Criticality Safety

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Rechargeable Li-ion battery with halogen intercalated graphite electrode

The disclosure provides rechargeable lithium ion batteries comprising at least one lithium salt-graphite composite electrode. In particular, the disclosure provides a rechargeable “water-in-bisalt” lithium ion battery with a high potential where at least a portion of the lithium salt is phase separated from the aqueous electrolyte, and where the anionic-redox reaction occurs within the graphitic lattice.

Wang, Chunsheng↗
DOE OSTI2026

NuMI Graphite Target Fin Removal Assembly

Materials scientists want to study the irradiated graphite target from the Neutrino at the Main Injector (NuMI) beam line. In order to gather these irradiated graphite fins from the target canister, it will have to be done remotely. Therefore, a removal assembly will need to be created in order to do this efficiently and well. This project aims to make progressive steps towards deploying the final target removal assembly. Furthermore, calculations related to the removal assembly design were done, and tests were planned to discover further information about the feasibility of this design. The main result from this project is the deflection of the cantilever arm, which is 0.354 inches due to the weight of the beam and cutting tool. Additionally, more results of this project are the testing assembly CAD model and the feasibility of design. This project concludes that there is still more work that needs to be done for the removal assembly to be deployed, but there were foundational steps that were made in the implementation of this design. Finally, there should be further steps and tests to determine the feasibility of this removal assembly.

Malovski, Azra [NIU, DeKalb]↗
DOE OSTI2023

Qualifying nuclear graphite components using ASME guidelines

Qualifying nuclear graphite components using ASME guidelines, including Semi-probabilistic, probabilistic, and deterministic approaches to design, theory of methods, pre-assessment analysis inputs, post assessment analysis outputs, simplified assessment, full assessment, and application.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Evaluating temperature dependent degradation mechanisms of silicon-graphite electrodes and the effect of fluoroethylene carbonate electrolyte additive

Understanding and overcoming the relatively high rate of degradation observed with silicon (Si)-based anodes in lithium (Li)-ion batteries (LiBs) is crucial for developing cells with increased energy density. The capacity fade of Si-graphite (Gr) electrodes in electrolytes with and without 10 wt% fluoroethylene carbonate (FEC) is explored during cycling at temperatures of 25, 45, and 70 °C, focusing on the complex degradation modes arising from loss of active material, loss of Li inventory, and cell resistance growth. We first show the common half-cell testing cannot be used at elevated temperatures as the performance of Li metal overshadows the behavior of Si-Gr electrode. Symmetric cell configuration successfully eliminates the effect of foreign materials, and reveals that while the presence of FEC provides better passivation of Si, it also results in greater parasitic reactions at 70 °C and faster capacity fade. The electrolyte without FEC, on the other hand, shows higher capacity retention at 70 °C. Finally, we propose a multi-step test protocol which can successfully deconvolute the information on loss of Li inventory and of active material. By combining half-cell and symmetric cell test methods, we demonstrate that loss of active material is reduced at higher temperatures and the dominant mechanisms of capacity loss for Si-Gr electrodes at elevated temperatures is loss of Li inventory.

25 ENERGY STORAGE↗
DOE OSTI2025

Development of the technological process for the IGR reactor's highly-enriched irradiated uranium-graphite fuel immobilization

The immobilization of irradiated highly enriched uranium (HEU) fuel is a critical component of nuclear waste management and non-proliferation efforts. In Kazakhstan, at National Nuclear Center of the Republic of Kazakhstan special attention is given to managing legacy HEU fuel from research reactors. One such case involves the IGR research reactor, whose first core containing irradiated HEU uranium-graphite fuel was operated from 1961 to 1966 and removed following reactor modernization. This fuel now requires a reliable and secure immobilization strategy. Here, this paper presents the development of a technological process for immobilizing this fuel to reduce its enrichment to below 20 % in terms of 235U content. The proposed method involves down-blending irradiated HEU fuel with depleted uranium, followed by encapsulation in a Portland cement matrix. Full-scale experiments were conducted to assess the uniformity of uranium distribution within the matrix. The results confirm the effectiveness of this approach, ensuring reliable immobilization of fuel in accordance with international requirements, including IAEA standards and Kazakhstan's regulatory framework. These findings contribute to the broader effort of adapting immobilization strategies for the safe management of spent fuel from research reactors.

12 - MGMT OF RADIOACTIVE AND NON-RADIOACTIVE WASTE↗
DOE OSTI2023

Structure–property relations in graphitic pebbles for nuclear applications

This work presents an analytical approach for holistically characterizing graphitic matrix pebbles for nuclear applications whereby the macrostructure, microstructure, and thermophysical properties of pebbles are determined. A systematic sectioning method was applied to several pebbles to describe the regional properties of the samples. Intact matrix-only spheres and sections of spheres fabricated by Kairos Power were characterized via optical imaging, x-ray computed tomography, x-ray diffraction, and ellipsometry to determine 2D and 3D macrostructure and anisotropy. The thermophysical properties of these materials were determined via measurements of density, specific heat, thermal expansion, and thermal diffusivity. The results of this study indicate that the pebble fabrication methods and their resultant effect on microstructure have a nontrivial effect on thermophysical properties, confirming the importance of robust characterization of these components. A discussion of the characterization approach and its applicability to nuclear fuel development activities is also included.

42 ENGINEERING↗
DOE OSTI2024

Halide-free synthesis of metastable graphitic BC 3

A halide-free route to the synthesis of graphitic BC 3 was discovered via decomposition temperature matching: benzene (C 6 H 6 ) as the carbon precursor and the borohydride anion (BH 4 − ) as the boron precursor.

08 HYDROGEN↗
DOE OSTI2020

Detonation-induced transformation of graphite to hexagonal diamond

In this work, we explore the structural evolution of highly oriented pyrolytic graphite (HOPG) under detonation-induced shock conditions using in situ synchrotron x-ray diffraction in the ns timescale. We observe the formation of hexagonal diamond (lonsdaleite) at pressures above 50 GPa, in qualitative agreement with recent gas gun experiments. First-principles density functional calculations reveal that under uniaxial compression, the energy barrier for the transition toward hexagonal diamond is lower than that for cubic diamond. Finally, no indication of cubic diamond formation was observed up to >70 GPa.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Coal-Tar-Pitch to Battery-Grade Graphite

Presentation for Domestic Production of Synthetic Graphite Roundtable, University of Kentucky Center for Applied Energy Research (CAER), Lexington, Kentucky.

01 COAL, LIGNITE, AND PEAT↗
DOE OSTI2020

Modeling Graphite Behavior (Fracture and Oxidation)

Purpose of this work Oxidation modeling Oxidation of full-scale components Local properties in oxidized components Graphite fracture strength Phase field fracture modeling Microstructurally informed fracture model IG-110 tensile results and comparison to experiment Work in progress and future work

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

ART Advance Graphite Creep (AGC) Irradiation Experiment

ART Advance Graphite Creep (AGC) Irradiation Experiment topics of discussion include: Schedule, AGC Experiment Update, AGC-4 Status, Anticipated areas data will be used, and Vendor specific irradiation capsule. This includes NRC/Licensing questions on irradiation behavior, behavior model development, and other collaborations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Graphite Oxidation Activities

Graphite Oxidation Activities to include introduction, rate determination, strength after oxidation, penetration depth analysis, strategic partnership projects, and summary and continuing work.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗